ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller & Associates I MJW Corporation

Page 1 of 24

Document Title:

Estimating the Maximum Plausible Dose 

to Workers at Atomic Weapons Employer 

Facilities

Document Number: ORAUT-OTIB-0004

Revision: 03 PC-1

Effective Date: 11/18/2005

Type of Document: OTIB

Supersedes: Revision 02

Subject Experts: Jeri L. Anderson and Cindy W. Bloom

Document Owner

Approval: Signature on File Approval Date: 08/02/2005

Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 08/04/2005

Richard E. Toohey, Project Director

Approval: Signature on File Approval Date: 08/12/2005

James W. Neton, Associate Director for Science

 New Total Rewrite Revision Page Change

DOCUMENTS MARKED AS A TOTAL REWRITE, REVISION, OR PAGE CHANGE REPLACE THE PRIOR REVISION. 

PLEASE DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-OTIB-0004 Revision No. 03 PC-1 Effective Date: 11/18/2005 Page 2 of 24

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION

NUMBER DESCRIPTION

11/20/2003 00 New Technical Information Bulletin to establish a process for 

estimating the maximum plausible dose to workers at selected AWE 

Facilities. First approved issue. Initiated by Jeri L. Anderson.

11/25/2003 01 Approved issue of revision 01. Initiated by Jeri L. Anderson.

12/04/2003 02 Approved issue of revision 02. Initiated by Jeri L. Anderson

08/12/2005 03 Approved issue of Revision 03. Training required: As determined by 

the Task Manager. Initiated by Cindy W. Bloom.

11/18/2005 03 PC-1 Approved page change revision to correct enriched uranium intake 

numbers in Table 3-5 (page 12). No sections were deleted Training 

required: As determined by the Task Manager. Initiated by Cindy W. 

Bloom.

Approval:

Judson L. Kenoyer, Task 3 Manager

Signature on File________________________ __ __11/16/2005

Signature on File ______ Kate Kimpan, Project Director11/15/2005

James W. Neton, Associate Director for Science

Signature on File__________________________ __ _11/18/2005
Document No. ORAUT-OTIB-0004 Revision No. 03 Effective Date: 08/12/2005 Page 3 of 24

TABLE OF CONTENTS

Section

Publication Record ..............................................................................................................................

Page

2

1.0 Scope .........................................................................................................................................5

2.0 General Information ....................................................................................................................5

3.0 Uranium Metal Handling Facilities...............................................................................................6

3.1 Uranium Metal Handling Facilities – Estimation of Internal Dose........................................... 7

3.1.1 Uranium Inhalation.........................................................................................................8

3.1.2 Uranium Ingestion..........................................................................................................9

3.1.3 Consideration of Uranium Bioassay Data....................................................................... 9

3.1.4 Internal Dose: Recycled Uranium at Uranium Metal Handling Facilities ........................ 9

3.1.5 Internal dose: Depleted or Enriched Uranium.............................................................. 10

3.1.6 Occupational Internal Dose Reconstruction Assumptions and Summary ..................... 11

3.2 Uranium Metal Handling Facilities – Estimation of External Dose........................................ 12

3.2.1 Uranium Metal Handling Facilities – Penetrating External Dose................................... 13

3.2.2 Uranium Metal Handling Facilities – Nonpenetrating External Dose............................. 14

3.2.3 Occupationally Required Medical X-Ray at Uranium or Thorium Metal 

Handling Facilities........................................................................................................ 15

3.2.4 Occupational External Dose Reconstruction Assumptions and Summary .................... 16

3.3 Uranium Metal Handling Facilities – Estimation of Exposure to Residual 

Radioactivity........................................................................................................................16

References ........................................................................................................................................18

Attachment A Industrial Radiation Doses To Be Included in Maximum Estimates.......................... 20

Attachment B Uranium Metal Handling ..........................................................................................22
Document No. ORAUT-OTIB-0004 Revision No. 03 Effective Date: 08/12/2005 Page 4 of 24

LIST OF TABLES

Table

3-1 Assumed fractions of recycled uranium contaminants at metal handling 

facilities ....................................................................................................................................

Page

6

3-2 Predicted uranium urinalyses from maximum inhalation and ingestion 

chronic intakes .......................................................................................................................10

3-3 Estimated recycled uranium activity fractions for internal dose 

reconstruction ........................................................................................................................10

3-4 Uranium intake multiplication factor based on enrichment...................................................... 11

3-5 Internal exposure summary for operational period.................................................................. 12

3-6 Maximum size of the uranium ingots used in proceessing...................................................... 13

3-7 Calculated photon exposure rates from uranium metal ingots ................................................ 14

3-8 Internal exposure summary.................................................................................................... 16

A-1 Annual industrial dose distributions ........................................................................................ 21

B-1 List of applicable facilities....................................................................................................... 22
Document No. ORAUT-OTIB-0004 Revision No. 03 Effective Date: 08/12/2005 Page 5 of 24

1.0

Technical Information Bulletins (TIBs) include general working documents that provide guidance 

concerning the preparation of dose reconstructions at particular sites or categories of sites. TIBs will 

be revised in the event additional relevant information is obtained. TIBs may be used to assist NIOSH 

in the completion of individual dose reconstructions.

SCOPE

The purpose of this document is to provide guidance for estimating the maximum plausible dose to 

workers at Atomic Weapons Employers (AWEs). This document may also be used to estimate doses 

at Department of Energy facilities when the exposures would be adequately estimated by the methods 

in this document. (The classification of facilities as either DOE or AWE is based on regulatory 

definitions, not on the type or amount of radiation exposure that might have occurred). This document 

describes an efficiency process that may be used to expedite the processing of claims requiring dose 

reconstruction under the Energy Employees Occupational Illness Compensation Program Act 

(EEOICPA). The exposure matrices in this document are designed to estimate the maximum 

plausible annual doses for the purpose of expediting certain types of claims.

In this document, the word “facility” is used as a general term for an area, building or group of 

buildings that served a specific purpose at a site. It does not necessarily connote an “atomic weapons 

employer facility” or a “Department of Energy facility” as defined in the Energy Employees

Occupational Illness Compensation Program Act of 2000 (42 U.S.C. § 73841 (5) and (12)).

This document includes general information regarding radiation dose reconstruction, followed by 

specific facility category information and by the type(s) of radiation exposure that might have occurred 

at a given facility category. The document was originally created to overestimate doses categorized, 

generally, as natural uranium handling facilities, and so this revision begins with uranium handling 

facilities.

2.0

The intake and dose rates in this exposure matrix are to be used in conjunction with the individual 

worker’s covered employment dates, date of birth, and date of cancer diagnosis. Rather than pro-rate 

intake and dose rates, it is acceptable to use exposure values for a full year, as this will overestimate 

dose. Because worker bioassay and dosimetry data, and workplace monitoring data cannot be 

ignored, the worker’s data and associated workplace monitoring data, if available, should be reviewed 

to ensure that a dose reconstruction based on this exposure matrix is an overestimate for a particular 

worker’s circumstances.

GENERAL INFORMATION

Some of the facilities have multiple periods of Atomic Energy Commission (AEC) operation (not 

including periods designated as remediation). In these cases, the maximum dose estimates in 

sections designated as X.1 (internal) and X.2 (external) are applied for each period of AEC operations

and only for these periods. Direction for estimating exposure during periods of normal (e.g., nonAEC) operation or remediation is provided in sections designated as X.3.

The term “day” in this document refers to a calendar day. The term “workday” is used to describe a 

day at work. It is assumed that there are 250 workdays per year and that a workday is 8 hours long:

this results in 2000 hours of work per year. The distinction between workday and calendar day is 

especially important when considering internal dose rates for use in calculating organ doses.

NIOSH has concluded that non-DOE sources of radiation exposure are to be reconstructed for the 

covered employment period. The NIOSH method for adding the non-DOE exposures is to assign the 
Document No. ORAUT-OTIB-0004 Revision No. 03 Effective Date: 08/12/2005 Page 6 of 24

dose distributions in Table A-1, to all AWE employees, unless the doses are lower in the facilityspecific data. Because this is an overestimating approach, and the use of analytical x-ray equipment, 

radioactive sources and materials, and radiography devices would not have been unusual in 

operational programs, the default is to include the dose from the estimated total industrial exposure

distributions from these activities in AWE overestimates. However, the industrial radiography 

component in Table A-1 can be excluded on a case-by-case basis, when additional information is 

sufficient to support exclusion (note that job title might not be sufficient to make this determination if 

there is no information about the location of the radiation source, the area radiation levels and the 

location of the employees). Note that these estimates of industrial radiation dose represent a 

reasonable estimate of likely doses for most workers, but any available claim-specific information 

should be considered when it is available for employees, who were working with or in the direct 

vicinity of industrial radiation sources and/or devices. 

3.0

There were approximately 102 AWE facilities that handled natural uranium in support of the atomic 

weapons program. Of these facilities, 92% operated for 10 years or less and 76% operated for 5 

years or less. The maximum time operated was twenty-three years and the minimum time was less 

than a year.

URANIUM METAL HANDLING FACILITIES

The uranium metalworking processes at these facilities included reduction and recasting, rolling, 

machining, extruding, and forging; fuel element fabrication; and uranium recovery from scrap. A large 

number of the facilities handled only uranium metal. Most of the facilities considered in this section

were privately owned and their work was done for the AEC or Manhattan Engineer District (MED) on a 

short-term basis and/or in addition to their normal commercial operations. Section 3 only addresses 

uranium, its short-lived progeny and possible contaminants from receipt of recycled uranium. The 

assumptions regarding recycled uranium contaminants are further discussed in other subsections, but 

the assumed mass (in parts per billion (ppb)) and activity fractions, based primarily on a review of 

Fernald and Hanford information, are shown here because of their significance to both internal and 

external dose reconstruction. The determination of activity fraction is based on depleted uranium, and 

results in overestimates of the activity fractions for facilities that used normal or enriched uranium with 

associated larger specific activities.

Table 3-1. Assumed fractions of recycled uranium contaminants at metal handling 

facilities.

Recycled uranium contaminant Pu-239 Np-237 Tc-99 Th-232 Th-228

Activity fraction of contaminant in uranium 0.00246 0.00182 0.379 2.73E-06 2.73E-06

Contaminant in ppb of uranium 10 1,040 9,000 10,000 *

*assumes same activity as Th-232

Exposure from processing of thorium, radium, uranium ores, or other radioactive materials (except as 

recycled uranium contaminants) is not considered in this section. Some uranium scrap recovery 

operations processed only uranium; however, a number of these operations also processed uranium 

long-lived progeny, such as thorium-230 and radium-226; some sites were processing small amounts

of plutonium or other radioactive contaminants in the waste.

To summarize, Section 3 provides an estimate of the maximum plausible dose from uranium and its 

short-lived progeny at uranium metal-handling facilities. This section includes consideration of 

radiation exposure from recycled uranium contaminants within the uranium metal matrix. This section 

does not include consideration of exposures from uranium ore processing, long-lived uranium 

progeny, or other radioactive metal processing, which might also have occurred at some facilities 
Document No. ORAUT-OTIB-0004 Revision No. 03 Effective Date: 08/12/2005 Page 7 of 24

during some periods. If a facility processed other forms of uranium or other radionuclides, additional 

exposure information must be considered when reconstructing a maximum plausible dose estimate. 

3.1 URANIUM METAL HANDLING FACILITIES – ESTIMATION OF INTERNAL DOSE

During World War II and the time of the Manhattan Engineer District (MED), the permissible level for 

natural uranium air dust concentration was set at 500 μg/m3 for insoluble uranium compounds and 

150 μg/m3 for soluble uranium compounds. After the war, the University of Rochester (Rochester, 

NY) recommended lowering the permissible level to 50 μg/m3 for all uranium compounds on the basis 

of chemical toxicity. This level was also stated as 70 disintegrations per minute per cubic meter (70 

dpm/m3

) for natural uranium and was based primarily on animal studies. The Medical Division of the 

AEC New York Operations Office (NYOO) thought that the “maximum permissible level” should be 

based on human data and was thus unknown. Therefore, the level of 50 μg/m3 was generally referred 

to as the “preferred level” (AEC 1949). Oftentimes the contractors of the AEC used the term 

“Maximum Allowable (air) Concentration” (MAC) interchangeably with the “preferred level” (PL) and

reported air-sampling results as multiples of the MAC (NLO 1952, AEC 1953). When considering air 

concentrations reported in multiples of the PL or MAC, the actual assumed value of the PL or MAC 

should be verified.

In 1949, the Medical Division of the NYOO published a report on the health hazards at seven facilities 

that produced and/or processed uranium for the AEC. These facilities included Mallinckrodt Chemical 

Works, Harshaw Chemical Company, Linde Air Products, Electro Metallurgical Company, Simonds 

Saw and Steel Company, Vulcan Crucible Company (later known as Aliquippa Forge), and Vitro 

Manufacturing Company. These facilities were the earliest MED- or AEC-constructed or employed, 

uranium processors and, with the exception of Vulcan Crucible, were very large operations in 

comparison to the other AWE uranium handling facilities. The AEC used the information on work 

tasks with measured air concentrations in breathing zones, general areas and process areas to 

determine average air concentrations weighted by exposure times and summed these time-weighted 

air concentrations to determine daily time-weighted average air concentrations by job categories. Up 

until the time of the 1949 MED report, surveys by the NYOO indicated that out of 648 exposed 

workers at these plants, 9% were exposed to uranium air concentrations greater than 125 MAC 

(greater than 6250 μg/m3

), 9% were exposed at 25-125 MAC (1250-6250 μg/m3), and 82% were 

exposed to less than 25 MAC (less than 1250 μg/m3

). As a result of the NYOO report, significant 

improvements were made in the operational conditions such as re-design of ventilation systems, 

enclosing some processes, and using remote controls (AEC 1949). By the end of 1949, exposure 

levels were significantly reduced at these larger plants even though production levels increased 

(Mason 1958).

In addition, AEC decided to discontinue long-term work at some facilities where resources were not 

available to improve radiation safety. Vulcan Crucible (Aliquippa Forge) was a steel mill that 

contracted with the AEC to roll uranium billets into rods on a part-time basis from 1948 to 1949. The 

contract (AEC Contract No. AT-(30-1)-407) stated that the plant was to arrange to spend “at least two 

consecutive weeks out of every five consecutive weeks” performing the AEC contract work. Four of 

the most exposed workers in the 1949 study were from this plant and the report stated that 

discussions were under way to shut down this operation (AEC 1949). Operations were shut down in 

late 1949 and the plant was decontaminated to 1950 standards (Adams and Payne 1992).

Early exposure studies were not very selective in terms of particle size (Stannard 1988), and could 

have under or over accounted for respirable radioactive material. Two early uranium facility particle 

size distribution studies measured particle size distributions. Medians ranged from 1.22 to 1.80 μm

mass median diameter during rolling (Spiegl et al. no date; probably 1949 or 1950, the year was 
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unspecified), and 3.1 to 4.9 μm [probably mass median diameter, but not specified] in an area of 

visible dust at an ore processing and scrap recovery facility (AEC 1958). When adjusted for density, 

these results are considered consistent with International Commission on Radiological Protection 

(ICRP) Publication 66 default parameters for particle deposition (ICRP 1994a), so dose 

reconstructions should assume ICRP Publication 66 defaults (including a 5 μm activity median 

aerodynamic diameter). In addition, the air concentration measurements represented the amount of 

uranium in the air where the workers were located, but not necessarily, what was taken into the lung 

or the body. At many of the AWE uranium metal facilities, the daily-weighted average exposures 

become significant overestimates of annual exposures when multiplied by 250 workdays per year. 

This is because the uranium operations were intermittent. Usually, AEC measured airborne 

contamination with alpha counters, so it is reasonable to use the activity rather than the mass 

designation for the MAC for estimation of the maximum daily intake.

3.1.1

Human and animal studies have indicated that oxides of uranium can be very insoluble in the lung 

(ICRP 1995), indicating absorption type S (0.1% and 99.9% with clearance half-times on the order of 

10 minutes and 7000 days, respectively). Other in vitro dissolution studies of compounds found at 

uranium facilities have shown that oxides of uranium exhibit moderate solubility (Eidson 1994; 

Heffernan et al. 2001) suggesting absorption type M (10% and 90% with clearance half-times on the 

order of 10 minutes and 140 days, respectively). In vitro dissolution tests on oxides produced from 

uranium metal during depleted uranium armor penetrator tests have indicated multicomponent 

dissolution rates, with 25% of uranium dissolving with a half-time of less than or equal to 0.14 days 

and 75% dissolving with a half-time of 180 days. Because there was no specific information on the 

solubility of aerosols produced at the AWE uranium metal processing facilities, this analysis assumed 

that both types M and S were available. The selection of absorption type should depend on the organ 

of interest. Dose reconstructions should assume International Commission on Radiological Protection 

(ICRP) Publication 66 default parameters for particle deposition (ICRP 1994).

URANIUM INHALATION

For the purpose of estimating maximum plausible dose, it is assumed that a worker received a 

constant exposure to a maximum level of 7000 dpm/m3 (100 MAC) during AEC operation. This 

document assumes that the intake for a year is derived from the worker being exposed to 100 MAC

for eight hours per day, five days per week, 50 weeks per year, whereas time-weighted average 

studies of even the larger plants like Mallinckrodt show that the majority of time workers were exposed 

at lower levels than this. No credit is taken for breaks or working at tasks where radioactive material 

intakes would be much lower or nonexistent. 

Although the information presented in Section 3.1, indicates that a small fraction of workers had daily 

time weighted exposures of greater than 100 MAC, the records also show that the AWE uranium 

metal handling facilities were not processing uranium on every workday or even every week. A 

review of some of the busier AWE uranium metal handling facilities, including Simonds Saw and 

Steel, and Bridgeport Brass (Adrian), support the 100 MAC maximum plausible exposure estimate for 

each year of AEC operation. In addition, a Y-12 coworker dose study indicates even at the very busy,

production-oriented Y-12 plant (which also handled highly enriched uranium) an estimated natural 

uranium exposure of 100 MAC would overestimate all type M predicted 95th percentile annual intakes 

and would be a reasonable estimate or overestimate for the 50th percentile type S intakes for most 

years (ORAUT 2005). 

The maximum annual uranium inhalation intake due to chronic exposure is estimated by multiplying 

the maximum air concentration of 7000 dpm/m3 (5000 μg/m3

) by the ICRP 66 recommended 

breathing rate of 1.2 m3

/h and by the assumed 2000 work-hours per year. This results in an annual 
Document No. ORAUT-OTIB-0004 Revision No. 03 Effective Date: 08/12/2005 Page 9 of 24

chronic inhalation intake of 1.68E+07 dpm (1.2E+07 μg), which is equal to a daily intake rate of 

46,000 dpm/day (3.3E+04 μg/day) For the purpose of calculating annual organ doses, the intake (in 

dpm) is assumed to be entirely U-234. For intake periods of less than one year, the intake rates 

shown in Table 3-5 should be adjusted as noted in Section 3.1.6, unless site- and/or claim-specific 

information are available to justify a total intake equal to less than this document assumed annual 

intake.

3.1.2

In the case where inhalation intakes are calculated from air concentrations, ingestion intakes are also 

to be considered. NIOSH (2004) indicates that the ingestion rate, in terms of dpm for an 8-hour 

workday, can be estimated by multiplying the air concentration in dpm per cubic meter by a factor of 

0.2, so the uranium ingestion rate based on an air concentration of 7000 dpm/m3 would be 1400 

dpm/workday. To adjust this to ingestion intake per calendar day, 1400 dpm/workday is multiplied by 

250 workdays per year and divided by 365 days per year, which equals 959 dpm/d. In accordance 

with NIOSH 2004, the same f1-value as used for inhalation dose calculations is to be used for 

ingestion dose calculations.

URANIUM INGESTION

3.1.3

The uranium fusion photofluorimetry urinalyses, which were the typical uranium bioassay method at 

the AWEs, were frequently performed by the University of Rochester and the AEC’s New York 

Operations Office and were similar to those performed at other AEC facilities. The default detection 

threshold for uranium urinalysis is assumed to be 10 μg/L based on a reported sensitivity of 5 to 10 

μg/L for uranium fluorimetry urinalysis in the early years (Wilson 1958). 

CONSIDERATION OF URANIUM BIOASSAY DATA

Predicted uranium urinalysis results were calculated for the last day of assumed chronic intake 

periods of 1, 2, 5, 10, 20 and 50 years, assuming the maximum estimated inhalation and ingestion 

intakes of natural uranium. The results are shown in Table 3-2. Type M uranium comes to 90% 

equilibrium in urine at about one year resulting in excretion of about 1400 μg/L. Chronic inhalation 

and ingestion of Type S material results in a uranium urinalysis of about 50 μg/L after 1 year of 

exposure and about 160 μg/L after 50 years of exposure. A cursory review of uranium urinalysis data 

from the metal handling facilities (Simonds Saw and Steel, Bridgeport Brass (Havens, Adrian, 

Extrusion Plant and Seymour Specialty Wire), Chapman and Aliquippa Forge), as well as Y-12 and 

Paducah indicates that uranium urinalysis results exceeding 100 μg/L are rare and that most results 

are less than 50 μg/L. In addition, when results did exceed these values, they usually dropped to less 

than 30 μg/L within a week or so. This gives additional credibility to the assumption that the assumed 

uranium intake rates are likely to be overestimates for natural uranium. However, it should be noted 

that assumption of these intake rates for enriched uranium could result in much lower mass-based 

bioassay results. For 3.5% enrichment, the 1-year and 50-year bioassay results would be about 400 

μg/L for type M, and 16 and 49 μg/L, respectively, for type S. For 93.5% enrichment, the 1-year and 

50-year bioassay results would be about 14 μg/L for type M, and 0.5 and 1.6 μg/L, respectively, for 

type S.

3.1.4

Recycled uranium might have been processed at uranium metal handling facilities after 1952. An 

estimate of contaminants that might contribute the most to internal doses, based on a review of 

recycled uranium contaminants at Hanford and Fernald, is shown below in Table 3-2. (The relative 

INTERNAL DOSE: RECYCLED URANIUM AT URANIUM METAL HANDLING 

FACILITIES
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Table 3-2. Predicted uranium urinalyses from 

maximum inhalation and ingestion chronic intakes.

Bioassay 

collection year

Type M Type S

dpm/d μg/L* dpm/d μg/L*

0.5 2,579 1,215 84 39

1 2,909 1,370 108 51

2 3,066 1,444 146 69

5 3,113 1,466 210 99

10 3,138 1,479 255 120

20 3,162 1,490 295 139

50 3,185 1,501 333 157

*Mass results assume natural uranium exposure

internal dose contributions from Tc-99, Th-228 and Th-232 were low enough to be ignored.) It is 

unlikely that recycled uranium would constitute the entire source term nor would there be significant 

concentration of recycled uranium contaminants. In addition, the activity fractions assume that the 

uranium specific activity is based on depleted uranium, which increases the proportion of the 

contaminants by activity. The contaminant levels for depleted uranium overestimate the contaminants 

in uranium of normal enrichment by about 40%. The contaminants are assumed to be oxides, and it 

is assumed that a plutonium oxide might behave either as absorption type M or type S material. 

Neptunium is assumed to behave as absorption type M.

Table 3- 3. Estimated recycled 

uranium activity fractions for 

internal dose reconstruction.

Uranium Pu-239 Np-237

1 0.00246 0.00182

3.1.5

For a given uranium process, the mass of (long-lived) uranium released to air will not change because 

of enrichment. In other words, the mass concentration in air would likely be similar for depleted, 

natural and enriched uranium. However, as the enrichment of U-235 (and therefore U-234) increases, 

more activity would be in the air, because of an increase in uranium specific activity. (Specific activity 

refers to the radioactivity per unit mass of material; as the half-life increases, the specific activity 

decreases.)

INTERNAL DOSE: DEPLETED OR ENRICHED URANIUM

Depleting uranium results in reduced activity per mass, because the mass-ratio of the lower specific 

activity U-238 is increased. Because so few sites handled only depleted uranium, and because 

adjusting for depletion would reduce the estimated intakes below those estimated for natural uranium, 

it is assumed that depleted uranium intakes are adequately bounded by the natural or normal uranium 

intake overestimates.

Uranium enriched by gaseous diffusion or a calutron [an electromagnetic device for separating

material by mass] results in a significant increase in the shorter-lived U-235 and U-234 isotopes, 

which leads to an increase in specific activity of the uranium. Enriched uranium metal has essentially 

the same physical characteristics as natural or depleted uranium, but is more radioactive per unit 

mass of uranium. 

One way to estimate intakes from processing enriched uranium, when information is available 

regarding intakes of natural uranium from the same process, is to consider the different specific 

activities of the material being processed. The specific activities for various uranium compositions are 
Document No. ORAUT-OTIB-0004 Revision No. 03 Effective Date: 08/12/2005 Page 11 of 24

shown in Table 3-4. Assuming the amount of material that will become airborne is based on the 

physical and chemical characteristics of uranium, and is not related to its specific activity, it is 

reasonable to assume that mass-based air concentrations for a given process will be the same for 

natural and enriched uranium. It follows that an estimate of radioactive air concentrations for enriched 

uranium based on natural uranium will increase by the ratio of the specific activity of the enriched 

uranium to the natural uranium. This estimate does not consider changes in handling practices and 

throughput, which were likely to have occurred during the handling of enriched uranium and which 

would likely reduce intakes from this overestimating method.

Table 3-4. Uranium intake multiplication factor based 

on enrichment.

Uranium

Specific activity

pCi/μg

Uranium activity 

multiplier

Depleted 0.4021 1

Natural (normal) 0.6829 1

2% 1.616 2.37

3.5% 2.201 3.22

20% 9 13.2

93.5% 68.11 99.7

Intake activity multipliers are presented for various enrichments in Table 3-4. For sites where it is 

known that the fraction of enriched uranium processed is only a fraction of the total uranium 

processed, a weighted multiplier can be derived. If 80% of the uranium source term was natural, and 

20% was 93.5% enriched, the multiplication factor would be 0.8 times 1 plus 0.2 times 99.7, which 

equals 20.74. It is very unlikely that most of the AWEs handled large quantities of highly enriched 

(93.5%) uranium, because this would have required well-developed nuclear criticality safety 

programs. Even at low enrichments, the need to address nuclear safety issues would have deterred 

(but not eliminated) outsourcing of enriched uranium metalwork involving large quantities of material. 

For some sites, a more reasonable upper bound can be developed based on a review of the air 

monitoring data. If the air data appear to be sufficient to adequately bound the air concentrations 

during AWE operations, the results can be used to determine if the maximum natural uranium intake 

estimates in this document are bounding. If the natural uranium intake estimate bounds the air data 

and if bioassays do not exceed the activity-based predictions, no adjustment for enrichment is 

required for estimating maximum intakes.

3.1.6

Uranium intakes for overestimating internal dose at uranium metal handling facilities are shown in the 

table below. Multiple uranium intake possibilities are shown to allow the dose reconstructor to choose 

the uranium intake for the appropriate enrichment assumption. Note that only one uranium intake rate 

is to be assigned, in other words, the listed daily intakes of both natural uranium and enriched 

uranium should not be independently assigned. The type of uranium to assign depends on the 

assumptions regarding uranium isotopic composition at the facility. However, it is reasonable to 

weight multiple intake rates as noted in section 3.1.5. Appendix B provides facility-specific information 

for use in assigning intakes. Because this document provides an overestimating approach, and the 

choice of maximum plausible air concentration is based on time-weighting exposure for a year, 

caution should be used in applying this method to organ dose reconstruction for intake periods less 

than a year. For intake periods less than one year, a simple overestimating approach is to adjust the 

daily intake upward by the number of days in an average year, 365, divided by the number of days 

OCCUPATIONAL INTERNAL DOSE RECONSTRUCTION ASSUMPTIONS AND 

SUMMARY
Document No. ORAUT-OTIB-0004 Revision No. 03 PC-1 Effective Date: 11/18/2005 Page 12 of 24

exposed in that year. Reducing total intake to less than one year of intake, requires justification of the 

chosen intake rate for the particular site and employee circumstances. 

For dose calculations, uranium intakes are assumed to be U-234. Uranium oxides can be either 

absorption type M or S. 

Recycled uranium contaminants should be accounted for beginning in 1953 and every year thereafter. 

Neptunium oxides are type M. Plutonium oxides are assumed to be type M or S. 

Table 3-5 lists intake rate assumptions for overestimates of uranium metal handling facilities. The 

intake mode is chronic. The dose distribution is assumed to be constant. 

Table 3-5. Internal exposure summary for operational period. 

Radionuclide(s) Note Type

Intake 

mode

Intake

dpm/day

U-natural Only one of the listed 

uranium intakes is applied.

Assume uranium is U-234 

for dose calculations.

M, S Inhalation 4.60E+04

(a) Ingestion 9.59E+02

U (2%) M, S Inhalation 1.09E+05

(a) Ingestion 2.27E+03

U (3.5%) M, S Inhalation 1.48E+05

(a) Ingestion 3.09E+03

U (20%) M, S Inhalation 6.07E+05

(a) Ingestion 1.26E+04

U (93.5%) M, S Inhalation 4.59E+06

(a) Ingestion 9.56E+04

Pu-239 Do not include prior to 1953. M, S Inhalation 1.13E+02

(a) Ingestion 2.36E+00

Np-237 Do not include prior to 1953. M Inhalation 8.38E+01

(a) Ingestion 1.75E+00

a. Choose same f1-value as used for inhalation per NIOSH (2004).

3.2 URANIUM METAL HANDLING FACILITIES – ESTIMATION OF EXTERNAL DOSE

The majority of photons from natural uranium metals are in the 30 to 250 keV energy range. Solid 

uranium objects provide considerable shielding of the lower energy photons and harden the spectrum, 

causing the majority of photons emitted from a solid uranium object, such as a thick plate, to have 

energies greater than 250 keV. While it is recognized that solid uranium sources will have a hardened 

photon spectrum, exposure to a thin layer of uranium on a surface will result in a larger fraction of 

exposure to lower energy photons. This analysis assumed workers were exposed to photon energies 

in the 30 to 250 keV range, which is claimant-favorable when considering both organ dose conversion 

factors and radiation effectiveness factors. Nonpenetrating dose from natural uranium consists 

primarily of electrons with energies >15 keV. For consistent presentation, exposure or dose is 

reported as: 

• penetrating, assumed to be associated with photons of energies 30 keV or greater, and

• nonpenetrating, assumed to be associated with photons of energies less than 30 keV or with 

electrons.

After 1952, small quantities of primarily alpha (Pu-239, Np-237, and Th-232/228) and beta (Tc-99) 

emitting radionuclides found there way into the uranium metal via recycling. Because of their primarily 

nonpenetrating radiation types, relatively low activities, and relatively low external radiation hazard 
Document No. ORAUT-OTIB-0004 Revision No. 03 Effective Date: 08/12/2005 Page 13 of 24

when compared to that of their uranium matrix, their contribution to dose is considered adequately 

addressed by the uranium external dose estimates. A quick check of the relative penetrating dose 

from an overestimate of recycled uranium contaminants in ground surface contamination shows that 

they contribute less than 1% to penetrating dose. The assumed nonpenetrating doses from uranium 

in this document are sufficiently large to bound any small contribution from Tc-99.

Depleted uranium might have been used in small quantities at some facilities, but it is believed that 

the external dose assumptions are sufficiently large in this document to bound any slight differences 

in dose contributions from a small, depleted uranium source term. Some uranium metal handling 

facilities processed enriched uranium. Because of the nuclear safety concerns (criticality events), 

materials were usually stored away from the work area and throughput was reduced. In addition, 

enrichment of uranium reduces the activity ratio of U-238, as well as its progeny, which reduces both 

the photon and the beta exposure from the uranium metal. For enriched uranium, the additional 

photon dose contribution from U-235 and its progeny does not counteract this decrease. The 

assumptions for natural uranium are considered bounding for facilities that handled low enriched 

uranium metal. 

3.2.1

The maximum size of the uranium ingots used in the atomic weapons program was used to estimate 

external exposure during processing. The dimensions of the two larger ingots are shown below in 

Table 3-6.

Uranium Metal Handling Facilities – Penetrating External Dose

Table 3-6. Maximum size of the 

uranium ingots used in processing 

(Bonfer 2003).

Shape Dimensions (inches)

Rectangular 24 L × 16 W × 4 H

Cylindrical 20 L × 13 Dia.

It was assumed that the exposure geometry was 100% anterior-posterior (AP). The workers were 

assumed to spend 8 hours per day performing their duties and no credit was given for breaks or other 

periods when they might have been away from the work area. Although during performance of the 

job, the worker could have been any distance from the source (the ingots), to be claimant-favorable, it 

was assumed the worker spent the entire 8 hours at 1 foot from the source. 

For the calculations of exposure rate, it was assumed that the ingots were composed entirely of 

natural uranium metal that had aged for 100 days, which allows for the ingrowth of the short-lived 

progeny and adds significant dose contribution from the bremsstrahlung radiations due to beta 

emissions. More than half of the calculated penetrating dose is due to bremsstrahlung photons. The 

exposure rate at various distances from the ingots was calculated using MicroShieldTM (Grove 

Engineering 2003) and MCNP (LANL 2003). The calculated values are shown in Table 3-7.

The MCNP calculated dose rates were higher than the MicroShieldTM dose rates for both the 

rectangular and the cylindrical ingot. The calculated dose rate at 1 foot was higher from the 

rectangular ingot, so that value was used in the dose estimation. The MCNP-calculated dose rate 

was multiplied by 2000 hours per year, resulting in an annual dose of 4.16 rem. The organ doses are 

calculated by multiplying the calculated annual dose rate by the mode of the “Ambient Dose 

Equivalent (H*(10)) to Organ Dose (HT)”, AP, 30 -250 keV photon dose conversion factors in 

Appendix B of the NIOSH External Dose Reconstruction Implementation Guideline (NIOSH 2002). 
Document No. ORAUT-OTIB-0004 Revision No. 03 Effective Date: 08/12/2005 Page 14 of 24

Table 3-7. Calculated photon exposure rates from uranium metal 

ingots.

Shape of ingot

Distance 

from surface

MicroShieldTM

exposure 

rate (mR/hr)

MCNP dose 

rate (mrem/hr)

Rectangular 1 cm* 7.90 8.26

1 ft 1.91 2.08

1 m 0.29 0.373

Cylindrical 1 cm* 8.27 8.44

1 ft 1.13 1.15

1 m 0.17 0.185

*The MCNP dose rate is calculated at the surface of the ingot.

The chronic external exposure scenario is only a surrogate to estimate maximum doses, and is really 

a replacement for individual dosimetry measurements. Annual photon dose based on the sum of 

individual dosimetry measurements is assumed for probability of causation determination to occur 

acutely rather than chronically for a given year. This same claimant-favorable assumption is applied 

here. The annual organ doses from this scenario are entered into the NIOSH IREP program 

assuming acute dose from photons with energies from 30 to 250 keV and a constant dose distribution. 

It was also assumed that workers were exposed to contaminated surfaces during their entire work 

period. The level of surface contamination was determined by first calculating a terminal settling 

velocity for 5-μm activity mean aerodynamic diameter (AMAD) particles. The calculated terminal 

settling velocity of 0.00075 meters per second is multiplied by the assumed maximum air 

concentration of 7000 dpm/m3

. It was assumed that the surface contamination level was due to 365 

days of constant deposition from the constant air concentration. This is considered claimant-favorable 

as facility housekeeping practices most likely kept contamination to a lower level. At the assumed air 

concentration level with the calculated terminal settling velocity, a 365-day buildup will result in a 

contamination level of 1.66E+08 dpm/m2

. The annual organ dose due to exposure to contaminated 

surfaces was determined by multiplying the surface contamination level by the dose coefficients for 

contaminated ground surfaces from Federal Guidance Report No. 12 (EPA 1993). These EPA dose 

conversion factors were modeled using an anthropomorphic phantom standing on an isotropic plane 

source in air. Dose conversion factors, for organs other than skin, are based on photon exposure 

(including bremsstrahlung). Skin dose conversion factors include electron plus photon dose. The 

source term was assumed to be natural uranium contamination at the calculated level, plus its shortlived progeny in equilibrium. The organ doses from this scenario are entered into the NIOSH IREP 

program assuming an acute dose from photons with energies from 30 to 250 keV and a constant dose 

distribution.

Submersion penetrating doses were not considered, because the relative organ doses were more 

than a factor of 1000 smaller than the contaminated surface penetrating doses. 

Neutron doses can be considered negligible for natural and depleted uranium. For enriched uranium 

metal, neutron doses are likely to be much lower than photon doses. Intimately mixed uranium and 

low atomic number materials are unlikely to occur at the metal processing facilities, which further 

reduces the likelihood of significant neutron doses.

3.2.2

To estimate maximum nonpenetrating forearm and hand exposures, the maximum estimates of 

shallow dose in the Bethlehem Steel site profile (ORAUT 2004a) were used. Nonpenetrating 

exposures at Bethlehem Steel were based on the assumption that 60% of the time, a worker was 

Uranium Metal Handling Facilities – Nonpenetrating External Dose
Document No. ORAUT-OTIB-0004 Revision No. 03 Effective Date: 08/12/2005 Page 15 of 24

exposed to 0.15 rad/hour (1 foot from semi-infinite uranium source) and 40% of the time, a worker 

was exposed to 0.09 rad/hour (3 feet from semi-infinite uranium source), which results in an average 

dose rate of 0.126 rad/hr. For a 2000-hour work-year, this would result in a nonpenetrating dose of 

252 rad/year. The estimated annual dose of 252 rad is only included for dose reconstruction of skin 

on the hands and forearms. 

To put this into perspective, the annual skin dose at a depth of 7 mg/cm2 would be 230 rad for a 

worker who kept a portion of their skin in direct contact with a uranium slab (nonpenetrating dose rate 

at 7mg/cm2 is 0.230 rad/hour (ORAUT 2004a)) for 1000 hours (half the hours in a typical a work 

year). The contact shallow dose rate might be used as a starting value to determine other skin dose 

when claim-specific data indicates other areas of the skin were in close proximity to the uranium. For 

instance, when information indicates that workers sat on the rods and there is a possibly related skin 

cancer, it would be reasonable to estimate some fraction of time sitting on the rod, e.g., 30 minutes 

per workday, which would result in a “bare skin” exposure of 28.8 rad in a year. When clothing is 

taken into account, the garments would reduce the dose to the skin. To estimate the effect of 

clothing, it was assumed a 65% Dacron/35% cotton lab coat plus a white cotton glove were 

representative of the fabric thickness: their uranium reduction factors would be 0.91 and 0.89, 

respectively (DOE 2004), for a total reduction factor of 0.81, which would reduce the annual skin dose 

to 23.3 rad in a year. 

The estimated dose from nonpenetrating radiation impinging on areas of the body other than the 

hands and forearms is estimated as 10 times the 1-foot photon dose. This estimate is based on a 

preliminary review of film badge data that indicates the nonpenetrating dose is usually no more than a 

factor of 10 larger than the penetrating dose. This assumption results in maximum nonpenetrating 

dose of 41.6 rem in a year to areas of the body other than hands and forearms. Nonpenetrating 

doses are reconstructed for tissues and organs near the body surface, including breast, penis, testes, 

skin and eye, by applying an organ dose conversion factor. 

It should be noted that the beta (0.07 mg/cm2) dose rate at the surface of an aged uranium slab is 

approximately 0.23 rad per hour. Melting and casting can result in increases of beta dose rates to a

range of 1 to 20 rad per hour, because these operations cause the short-lived uranium progeny to 

concentrate at the surface of the metal (DOE 2004). The dose rates decrease as the short-lived 

progeny decay or are physically removed. Because this phenomena has been recognized for some 

time, it is likely that extra care was taken to minimize dose after uranium was just melted or cast, but 

dose reconstructors should be aware of these higher dose operations as they review individual 

claims. 

3.2.3

Information regarding whether or not occupationally required medical X-ray examinations were 

performed at AWEs is not always available. AEC usually, but not always, required “preemployment”, 

periodic (annual), and termination medical examinations of workers involved in the larger uranium and 

thorium programs. The term “preemployment” as used here, means prior to performing AECcontracted radiological work. Although a typical examination most likely involved a PA radiography 

chest examination using 14 x 17 inch films, photofluorography examinations might also have been 

employed. ORAUT-PROC-0061, Occupational X-Ray Dose Reconstruction for DOE Sites (ORAUT 

2004b), Table 1 shows that photofluorography examinations might have continued through 1961 at 

some DOE facilities. For the maximizing approach, occupationally required medical x-ray 

examinations are assumed to consist of photofluorography chest examinations through 1961 at the 

metal handling facilities. Beginning in 1962, required x-ray examinations are assumed to consist of 

Occupationally Required Medical X-Ray at Uranium or Thorium Metal Handling 

Facilities
Document No. ORAUT-OTIB-0004 Revision No. 03 Effective Date: 08/12/2005 Page 16 of 24

PA chest x-rays. Organ doses can be obtained or assigned in accordance with information from the 

current revision of ORAUT-OTIB-0006, Technical Information Bulletin: Dose Reconstruction from 

Occupationally Related Diagnostic X-Ray Procedures (ORAUT 2003).

3.2.4

Table 3-8 summarized the components of external dose reconstruction for both operational and 

residual exposure periods. The basis for the residual exposure assumptions are presented in the next 

section 3.2.5. To expedite and simplify inclusion of organ doses due to industrial radiation sources, it 

is reasonable and claimant-favorable to use the 1939 total dose distribution for all years adjusted by

the mode of the “Ambient Dose Equivalent (H*(10)) to Organ Dose (HT)”, AP, 30 -250 keV photon 

dose conversion factors in Appendix B of the NIOSH External Dose Reconstruction Implementation 

Guideline (NIOSH 2002), until a more efficient method is available for inclusion of these doses. 

Occupational External Dose Reconstruction Assumptions and Summary

Table 3-8. External exposure summary.

Source Exposure Information Notes

Period of 

applicability Annual exposure

IREP 

distribution

Uranium metal Penetrating (Photons, 30-

250 keV, AP, acute)

Operational 

covered period

4.16 rem Constant

Uranium 

contamination

Penetrating (Photons, 30-

250 keV, AP, acute)

Accounts for skin dose and 

penetrating dose from 

surface contamination.

Does not account for

nonpenetrating dose to 

other shallow organs.

Operational 

covered period 

and residual 

period

Radiation from 1.66E+08 

dpm/m2of natural 

uranium surface 

contamination, plus 

equilibrium 

contamination amounts 

of short-lived progeny 

Constant

Uranium metal Nonpenetrating (Electrons 

>15 keV, acute)

Skin on hands and 

forearms

Operational 

covered period

252 rem Constant

Uranium metal Nonpenetrating

(Electrons >15 keV, acute)

Skin (other than on hands 

and forearms) and shallow 

organs

Operational 

covered period

41.6 rem Constant

Uranium 

contamination

Nonpenetrating

(Electrons >15 keV, acute)

Shallow organs other than 

skin

Residual period 10 times the penetrating 

dose derived from 

uranium contamination

Constant

Industrial 

Radiography

Penetrating (Photons, 30-

250 keV, AP, acute)

Operational 

covered period

See Appendix A Lognormal

Other industrial 

sources

Penetrating (Photons, 30-

250 keV, AP, acute)

Operational 

covered period

See Appendix A Lognormal

Medical X-ray Photofluorographic chest 

examination

Initial plus one examination 

per year plus termination

Before 1962 See ORAU 2003

Medical X-ray PA radiographic chest 

examination

Initial plus one examination 

per year plus termination

After 1961 See ORAU 2003

3.3 URANIUM METAL HANDLING FACILITIES – ESTIMATION OF EXPOSURE TO 

RESIDUAL RADIOACTIVITY

During operations with uranium, airborne contamination can be caused directly from operations with 

uranium bearing materials as well as by the resuspension of surface contamination. This airborne 

contamination can be removed by settling on floors or other surfaces or by ventilation (both 

engineered and natural building ventilation). The surface contamination levels can be increased by 

the airborne contamination settling or directly by operations with uranium. This level is decreased by 

various incidental means such as general housekeeping or tracking of contamination out of the area 

on equipment or personnel.

For the purpose of estimating the airborne contamination from residual contamination, it was assumed 

that during uranium operations, these factors reached an equilibrium that caused the airborne 

concentration and surface contamination to remain at a constant level. Once the generation source 
Document No. ORAUT-OTIB-0004 Revision No. 03 Effective Date: 08/12/2005 Page 17 of 24

from uranium operation was removed, the other factors then caused the over mass of uranium to 

decrease at some rate due to the remaining removal factors.

This analysis assumes that 1) uranium operations directly accounted for half the airborne 

concentration in the facility during operations and 2) all the remaining removal mechanisms eliminated 

only 1% of the uranium mass per day. With these assumptions in place, an equation can be written 

for the airborne concentration as a function of time.

t C t MAC e 0.01* ( ) 50 * − =

Where:

C(t) = Airborne activity at time t (pCi/m3

)

t = time since uranium operations ended (days)

MAC = Maximum Allowed Concentration (70 dpm/m3

)

0.01 = fraction of material removed per workday

When this is multiplied by the breathing rate of 9.6 m3 per workday, the daily intake is calculated. This 

equation can then be integrated from the end of uranium operations to infinity to determine the total 

amount of uranium that could be inhaled from residual contamination. This integration results in:

I = 9.6 * 50MAC / 0.01

Where:

I = total intake from inhalation of residual activity (dpm)

9.6 = daily breathing rate (m3

/workday)

A total uranium inhalation intake of 3.36E+6 dpm was assumed to occur during the residual 

contamination period. This equates to 20% of the 1.68E+7 pCi annual inhalation intake during 

operations. Therefore, the total uranium inhalation intake from residual contamination can be 

overestimated by assuming one additional year of inhalation exposure at the operational level of 100 

MAC. In addition, one additional year of intake of the recycled uranium contaminants would also be 

assumed. The intake rates for airborne exposures at 100 MAC are listed in Table 3-5.

Ingestion of material is most closely related to removable surface contamination rather than fixed 

surface contamination. Airborne concentrations are also related to removable surface contamination. 

Therefore, the removable contamination levels are assumed to decrease at the same rate as the 

airborne concentrations. For this reason, the same approach will be taken for ingestion estimates 

following periods of operations. That is, the ingestion intakes from residual contamination will be 

estimated by assuming one additional year of ingestion at the same rate it occurred during operation.

The external radiation caused by residual contamination can be overestimated by assuming the 

contamination levels that existed during operations continued to exist after operations ceased. The 

method for estimating the annual dose from this source of radiation is specified in section 3.2.1 for

penetrating dose and for nonpenetrating skin dose. Nonpenetrating annual organ doses for the 

breast and testes can be estimated by multiplying their annual penetrating doses determined by this 

method by a factor of ten. These external doses should be added to an individual’s estimate for each 

year the individual worked at the facility following the end of uranium operations, during periods of 

DOE remediation, and during breaks between AEC operations in the case of multiple operational 

periods.
Document No. ORAUT-OTIB-0004 Revision No. 03 Effective Date: 08/12/2005 Page 18 of 24

REFERENCES

Adams, W.C., and Payne, J.L., 1992, “Characterization Survey of Portions of the Aliquippa Forge Site, 

West Aliquippa, Pennsylvania,” prepared for the Office of Environmental Restoration, U.S. 

Department of Energy, ORISE 93/A-6, December. 

AEC (U.S. Atomic Energy Commission), 1949, “Health Hazards in NYOO Facilities Producing and 

Processing Uranium: A Status Report- April 1, 1949,” Prepared by the New York Operations 

Office Medical Division, Issued April 18 [Mallinckrodt\08-28-03 – Data Capture\010001510].

AEC (U.S. Atomic Energy Commission), 1953, “Simonds Saw and Steel Company, Occupational 

Exposure to Uranium,” New York Operations Office Health and Safety Division, February 18

[pp. 530-535 of 685 of Simonds Saw and Steel Company.pdf]. 

AEC (U.S. Atomic Energy Commission), 1958, “Kerr McGee Oil Industries, Inc., Shiprock, New 

Mexico, Occupational and Non-Occupational Radiation and Contamination,” HASL-18, CM-6, 

New York Operations Office Health and Safety Laboratory, February 7. 

[EmlBx27KerrMcgeeOilIndIncShiprockNMFeb1958] 

Bonfer, D., 2003, Personal Communication with J. L. Anderson.

DOE (U.S. Department of Energy), 2004, Guide of Good Practices for Occupational Radiological 

Protection in Uranium Facilities, DOE-STD-1136-2004, December.

Eheman, C.R. and P. E. Tolbert, 1999, “Estimating Occupational Radiation Doses When Individual 

Dosimetry Information Is Not Available: a Job Exposure Matrix. American Journal of Industrial 

Medicine,” 36: 348-359.

Eidson, A. F., 1994, “The effect of solubility on inhaled uranium compound clearance: A review,” 

Health Physics, volume 67, number 1, pp. 1–14.

EPA (US Environmental Protection Agency), 1984, “Occupational Exposure to Ionizing Radiation in 

the United States: A Comprehensive Review for the Year 1980 and a Summary of Trends for 

the Years 1960-1985,” EPA 520/1-84-005, Washington, D.C. 

EPA (US Environmental Protection Agency), 1993, External Exposure to Radionuclides in Air, Water, 

and Soil, Federal Guidance Report No. 12, September.

Grove Engineering, MicroShield Software Program, Version 6.01, Copyright 1992-2003.

Heffernan, T. E., J. C. Lodwick, H. Spitz, J. Neton, and M. Soldano, 2001, “Solubility of airborne 

uranium compounds at the Fernald Environmental Management Project,” Health Physics, 

volume 80, number 3, pp. 255–262.

International Commission on Radiological Protection (ICRP), 1994, Human Respiratory Tract Model 

for Radiological Protection, ICRP Publication 66, p. 198.

International Commission on Radiological Protection (ICRP), 1995, Age-dependent Doses to 

Members of the Public from Intake of Radionuclides: Part 4 Inhalation Dose Coefficients, ICRP 

Publication 71, p. 299.
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LANL, X-5 Monte Carlo Team, 2003, “MCNP—A General Monte Carlo N-Particle Transport Code, 

Version 5,” LA-UR-03-1987, Los Alamos National Laboratory, April 24.

Mason, M.G., “A Summary of Fifteen Years of Experience With Dust Problems in Refining and 

Fabrication of Uranium,” in proceedings of the Symposium on Occupational Health Experience 

and Practices in the Uranium Industry, HASL-58, October 15-17, 1958.

National Lead Company of Ohio (NLO), 1952, Production Report on the Rolling of Two Hundred and 

Eighteen Uranium Billets at Bethlehem Steel Corporation’s Lackawanna Plant on March 15, 

1952, Contract No. AT(30-1)-1156, March 21.

NIOSH (National Institute for Occupational Safety and Health), 2002, External Dose Reconstruction 

Implementation Guideline, Revision 1, OCAS-IG-001, Office of Compensation Analysis and 

Support, Cincinnati, Ohio, August.

NIOSH (National Institute for Occupational Safety and Health), 2004, Estimation of Ingestion Intakes, 

Technical Information Bulletin, OCAS-TIB-009, Revision 0, Office of Compensation Analysis 

and Support, Cincinnati, Ohio, April 13.

ORAUT (Oak Ridge Associated Universities Team), 2003, Technical Information Bulletin: Dose 

Reconstruction from Occupationally Related Diagnostic X-Ray Procedures, 

ORAUT-OTIB-0006, Revision 02, Cincinnati, Ohio.

ORAUT (Oak Ridge Associated Universities Team), 2004a, Technical Basis Document: Basis for 

Development of an Exposure Matrix for Bethlehem Steel Corporation, Lackawanna, New York: 

Period of Operation: 1949-1952, ORAUT-TKBS-0001, Revision 01, Cincinnati, Ohio.

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Reconstruction Procedure, ORAUT-PROC-0061, Revision 00, Cincinnati, Ohio.

ORAUT (Oak Ridge Associated Universities Team), 2005, Technical Information Bulletin: Internal 

Dosimetry Coworker Data for Y-12, ORAUT-OTIB-0029, Revision 00, Cincinnati, Ohio.

Spiegl, C. J., H. E. Stokinger, J. DeVoldre, K. Lauterbach, and S. Dzuibs, no date, Air Contamination 

During Uranium Fabrication at Simonds Saw and Steel Company, S.M.3 (Toxic Limits), S.M. 

(Special Materials), January 12. [260000131.pdf]

Stannard, J.N., Radioactivity and Health: A History, Volume 1, Laboratory Research

Wilson, R. H., 1958, “The Hanford Uranium Bio-Assay Program,” Symposium on Occupational Health 

Experience and Practices in the Uranium Industry, HASL-58, U. S. Atomic Energy 

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, prepared for the 

U.S. Department of Energy, Office of Health and Environmental Research, DOE/RL/01830-

T59, October 1988.
Document No. ORAUT-OTIB-0004 Revision No. 03 Effective Date: 08/12/2005 Page 20 of 24

ATTACHMENT A

Industrial Radiation Doses To Be Included in Maximum Estimates

Table A-1 is a summary of annual external dose distributions for industrial users of radiation sources 

based on EPA (1984) summary dose data for four radiography periods and three other industrial 

exposure periods during 1960 through 1980, which were further analyzed as cumulative probability 

distributions (Eheman and Tolbert 1999). The dose data for the reported periods were assumed to 

occur in a year in the middle of the reported range. These dates and doses were fit to a line to 

estimate earlier doses. It was assumed that the doses did not change after 1979. The geometric 

standard deviations (GSDs) prior to 1960 were calculated by fitting the dates and product of the 

geometric means (GMs) and their GSDs, i.e., the 84th percentile doses, to a line; predicting the pre1960 years and than dividing those 84th percentile values by the geometric mean (assumed to be the 

median) to estimate GSDs. For each year of a worker’s covered employment period, the industrial 

radiography plus the other industrial exposure is to be included in the dose reconstruction, except as 

noted in Section 2.0.
Document No. ORAUT-OTIB-0004 Revision No. 03 Effective Date: 08/12/2005 Page 21 of 24

Table A-1. Annual industrial dose distributions

Year

Industrial radiography

exposure

Other industrial 

exposure

Total industrial 

exposure

GM rem GSD GM rem GSD GM rem GSD

1939 0.432 3.51 0.126 2.69 0.558 3.32

1940 0.424 3.5 0.125 2.67 0.548 3.31

1941 0.415 3.49 0.123 2.65 0.539 3.30

1942 0.407 3.48 0.122 2.64 0.529 3.29

1943 0.399 3.47 0.121 2.62 0.519 3.27

1944 0.390 3.46 0.119 2.6 0.509 3.26

1945 0.382 3.45 0.118 2.58 0.500 3.25

1946 0.374 3.44 0.116 2.56 0.490 3.23

1947 0.365 3.43 0.115 2.54 0.480 3.22

1948 0.357 3.41 0.113 2.52 0.470 3.20

1949 0.349 3.4 0.112 2.5 0.461 3.18

1950 0.340 3.39 0.110 2.48 0.451 3.17

1951 0.332 3.37 0.109 2.46 0.441 3.15

1952 0.324 3.36 0.108 2.44 0.431 3.13

1953 0.316 3.34 0.106 2.42 0.422 3.11

1954 0.307 3.33 0.105 2.39 0.412 3.09

1955 0.299 3.31 0.103 2.37 0.402 3.07

1956 0.291 3.29 0.102 2.34 0.393 3.04

1957 0.282 3.27 0.100 2.32 0.383 3.02

1958 0.274 3.25 0.099 2.29 0.373 3.00

1959 0.266 3.23 0.098 2.27 0.363 2.97

1960 0.250 3.13 0.091 2.12 0.341 2.86

1961 0.250 3.13 0.091 2.12 0.341 2.86

1962 0.250 3.13 0.091 2.12 0.341 2.86

1963 0.250 3.13 0.091 2.12 0.341 2.86

1964 0.250 3.13 0.091 2.12 0.341 2.86

1965 0.250 3.13 0.091 2.12 0.341 2.86

1966 0.171 2.94 0.091 2.12 0.262 2.66

1967 0.171 2.94 0.091 2.12 0.262 2.66

1968 0.171 2.94 0.091 2.12 0.262 2.66

1969 0.171 2.94 0.091 2.12 0.262 2.66

1970 0.171 2.94 0.069 1.53 0.240 2.54

1971 0.119 2.3 0.069 1.53 0.188 2.02

1972 0.119 2.3 0.069 1.53 0.188 2.02

1973 0.119 2.3 0.069 1.53 0.188 2.02

1974 0.119 2.3 0.069 1.53 0.188 2.02

1975 0.119 2.3 0.069 1.53 0.188 2.02

1976 0.119 2.3 0.069 1.53 0.188 2.02

1977 0.119 2.3 0.069 1.53 0.188 2.02

1978 0.119 2.3 0.069 1.53 0.188 2.02

1979 0.119 2.3 0.069 1.53 0.188 2.02

1980+ 0.107 2.1 0.071 1.56 0.178 1.88
Document No. ORAUT-OTIB-0004 Revision No. 03 Effective Date: 08/12/2005 Page 22 of 24

ATTACHMENT B

Uranium Metal Handling

Table B-1. List of applicable facilities. Site name Years of operation U enrichment Account for other radiation sources?*

Ajax Magnethermic Corp. 1958–1962 Normal No

Alba Craft 1952–1957 Normal No

Aliquippa Forge 1947–1950 Normal No

Allegheny-Ludlum Steel 1950–1952 Normal No

Aluminum Co. of America (NJ) 1944 Normal No

Aluminum Co. of America (PA) 1944–1945 Normal No

AMCOT 1961–1962 Normal No

American Bearing Corp. 1954–1959 Normal No

American Chain and Cable Co. 1944 Normal No

American Machine and Foundry 1951-1954 Normal Thorium

American Peddinghaus Corp. 1978 Normal No

Anaconda Co. (Connecticut) a 1942, 1956–1959 Normal No

Armco-Rustless Iron & Steel 1948 Normal No

Associated Aircraft Tool and 

Manufacturing

1956 Normal No

B&T Metals 1943 Normal No

Baker Brothers 1943–1944 Normal No

Baker-Perkins Co. 1956 Normal No

Bell Telephone Laboratoriesb 1943–1944 Normal No (small check sources not considered a 

health hazard)

Bendix Aviation (Pioneer Division) 1960 Normal No

Besley-Wells 1953 Normal No

Bethlehem Steel 1949–1952 Normal No

Birdsboro Steel & Foundryc 1951–1952, 1962 Normal No

Bliss & Laughlin Steel 1948–1952 Normal No

Bloomfield Tool Co. 1947, 1951 Normal No

Bridgeport Brass Co., Adrian 1954-1962 2% Thorium

Bridgeport Brass Co., Havens Lab Normal Thorium

Brush Beryllium Co. (Cleveland) 1942-1943, 1949-1953 Normalg Thorium

Brush Beryllium Co. (Detroit) 1940s-1950s Normal No

BWX Technologies (Virginia) 1959, 1968-1972, 1985-2001 93.5% Thorium, Plutonium, LINAC

C.H. Schnoor 1943–1951 Normal No

C.I. Hayes, Inc. 1964 Normal No

Callite Tungsten Co. 1944 Normal No

Carboloy Co.d 1956 3.5% k No

Carborundum 1944 (termination prior to 1960) Normal Plutonium 1960-1962

Carpenter Steel Co. 1943–1944 Normal No

Chambersburg Engineering Co. 1957 Normal No

Chapman Valve 1948–1949 Normal No

Cincinnati Milling Machine Co. 1963 Normal No

Combustion Engineering 1965–1972 93.5% No

Copperweld Steel 1943–1946 Normal No

Crane Company 1947-1949 Normal Thorium

Dow Chemical Co. (Madison Site) 1957–1960 Normal Thorium

Electro Circuits, Inc. 1952–1953 Normal No

Extruded Metals Company 1944 Normal No

Extrusion Plant 1962-present 3.5% k No

Fenn Machinery Co. 1950 Normal No

Fenwal, Inc. 1967–1968 0.95% No

Frankford Arsenal 1952–1954 Normalh No

Granite City Steel 1953–1966 Normal Radiography (betatron)

Great Lakes Carbon Workse 1952-1958 93.5% Thorium

Gruen Watch 1956 Normal No

Heald Machine Co. 1960 Normal No

Heppenstall Co. 1955 Normal No

Herring-Hall Marvin Safe Co. 1943–1951 Normal No
Document No. ORAUT-OTIB-0004 Revision No. 03 Effective Date: 08/12/2005 Page 23 of 24

Table B-1 (Continued). List of applicable facilities. Site name Years of operation U enrichment Account for other radiation sources?*

Hunter Douglas Aluminum Corp. 1959–1963 Normal Thorium

International Nickel Co., Bayonne 

Laboratories

1951–1952 Normal No

International Register 1943 Normal No

Ithaca Gun Co. 1961–1962 Normal No

J.T. Baker Chemical Co. 1948, 1957–1958 Normal

Jessop Steel Co. 1950–1954 Normal No

Joslyn Manufacturing and Supply 

Co.

1944–1952 Normal No

Kaiser Aluminum Corp. 1959 Normal No

La Pointe Machine and Tool Co. 1956 Normal No

Landis Machine Tool Co. 1952 Normal No

Magnus Brass Co. 1954–1957 Normal No

McKinney Tool and Manufacturing 1944 Normal No

Medart Co. 1951–1952 Normal No

Midwest Machining 1944 Normal No

Mitchell Steel Co. 1954 Normal No

New York University 1946–1952 Normal No

Norton Company 1945-1957 Normal Thorium

Oliver Corp. 1956–1957, 1961–1962 3.5% k No

Podbeliniac Corp. 1957 Normal No

Precision Extrusion Co. 1949–1950, 1956–1959 Normal No

Quality Hardware and Machine 

Co.

1944–1945 Normal No

R.W. Leblond Machine Tool Co. 1961 Normal No

Reed Rolled Thread 1955 Normal No

Revere Copper and Brass 1943-1950s Normal No

Roger Iron Co. 1956 Normal No

Sciaky Brothers, Inc. 1953 Normal No

Seymour Specialty Wire 1962–1964 Normal No

Simonds Saw and Steel 1948-1956 Normali Thorium

Southern Research Institute 1955–1958, 1962, 1976 Normal No

Sperry Products, Inc. 1952–1953 Normal No

Star Cutter Corp. 1956 Normal No

Stauffer Metals, Inc. 1961 Normal No

Superior Steel Co. 1952–1957 Normali No

Sylvania Corning Nuclear Corp. –

Bayside

1947-1962 Normalj Thorium

Sylvania Corning Nuclear Corp. –

Hicksville Plant

1947-1962 3.5%j, k Thorium

Titus Metalsf 1956 93.5% No

Tocco Induction Heating Div. 1968–1969 Normal No

Torrington Co. 1951–1953 Normal No

Tube Reducing Co. 1952, 1957 Normal No

U.S. Steel Co., National Tube 

Division

1959–1960 Normal No

University of Michigan 1944 Normal No

Vulcan Tool Co. 1959 Normal No

W.E. Pratt and Company 1943-1946 Normal No

Wah Chang 1956-1979, 1971-1972 Normal Thorium

Westinghouse Atomic Power 

Development Plant

1941–1944 Normal No

Westinghouse Nuclear Fuels 

Division

1971-1972 3.5%k Plutonium

Wolverine Tube 1943-1946 Normal Thorium
Document No. ORAUT-OTIB-0004 Revision No. 03 Effective Date: 08/12/2005 Page 24 of 24

Table B-1 (Continued). List of applicable facilities. Site name Years of operation U enrichment Account for other radiation sources?*

Wyckoff Drawn Steel Co. 1943 Normal No

Wyckoff Steel Co. 1950 Normal No

(*Additional sources of radiation exposure have been identified for the listed site. An estimate of maximum plausible dose might be needed to account for 

radiation exposure from these other radiation sources. Updates will be made to this list as additional information becomes available.)

a. Anaconda was involved in gaseous diffusion barrier production in 1942. Based on available documentation, it appears that the later uranium work might 

have occurred in 1956 and in 1959, not 1956 through 1959, as noted by the DOE Office of Advocacy. Additional information is being sought to address 

this discrepancy. 

b. Bell Telephone Laboratories was involved in gaseous diffusion barrier production.

c. Concerns about radiography exposure were noted in at least one claim for this site. The described exposure might have occurred outside the covered 

period.

d. Their license indicated that enrichment up to 3% was allowed, 3.5% is used here because it is listed in Table 5.

e. TREAT reactor fuel, assumed HEU was 93.5%.

f. Argonaut reactor fuel, probably in the range of 20 to 90%, assumed 93.5%.

g. UF4 to metal production

h. UCl4

i. Although some EU was processed, normal uranium intakes will still provide an overestimate. 

j. Uranium hydrides

k. Enrichment assumptions are subject to change based on review of additional data, it is likely that the combination of enrichment and air concentration 

assumptions will overestimate intakes.